Localization Prototype Research of RCC-M Ejector in Nuclear Power Plant

2013 ◽  
Vol 421 ◽  
pp. 181-185
Author(s):  
Yan Li Qi ◽  
Yuan Xue ◽  
Tian Tian ◽  
Xing Jiang Chen ◽  
Xi Quan Fang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Stable Operation ◽  
Special Configuration ◽  
Safety Level ◽  
Practical Applications ◽  
Spray System ◽  
Nuclear Plants ◽  
Test Result

As the key equipment to ensure the stable operation of containment spray system in nuclear power plant, a RCC-M ejector with safe and credible performance is very important to make sure the containment integrality under lose of coolant accident (LOCA) or pipe broken. However, it was purchased from overseas supplier all along because of its special configuration, nonstandard design, severe safety level and small quantity demand. To implement equipment localization, and referring to practical applications of some nuclear plants, a RCC-M ejector is independently manufactured and tested, all performance could meet the requirement according to the test result. The RCC-M ejector is authenticated by experts by China Nuclear Energy Association, and the localization prototype research of this paper has great application and popularization value.

scholarly journals PiROB: Vision-based pipe-climbing robot for spray-pipe inspection in nuclear plants

2018 ◽  
Vol 15 (6) ◽  
pp. 172988141881797
Author(s):  
Jae-Hee Kim ◽  
Jae-Cheol Lee ◽  
You-Rack Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Mobile Robot ◽  
Nuclear Power ◽  
Climbing Robot ◽  
Pipe Inspection ◽  
Nuclear Plants ◽  
Reactor Building ◽  
Fail Safe ◽  
Very High

Among many pipes in a nuclear power plant, the spray pipes in the reactor building are one of the most important pipes in view of their function and safety aspects. However, it is very difficult to manually reach and inspect the pipes for defects or damage, because these pipes are installed in very high places. To carry out this kind of inspection more easily, we developed a mobile robot to climb up and down and to cross over such pipes. A mobile robot should be small and light enough that it can be practically and safely operated in a nuclear power plant. Our robot is able to overcome obstacles such as valves, pipe flanges, and T-shaped branches, and it also meets the requirements of fail-safe, autonomous grasping, and self-power without the cables to the remote control station. The robot has a five-degree-of-freedom manipulator and two grippers and moves along the cylindrical pipes bypassing the obstacles. The robot should be able to grasp the next pipe autonomously, because the robot works in places high off the ground where the remote operator cannot see the next pipe for the robot to grasp. This article proposes a vision-based scheme for grasping a cylindrical pipe semi-autonomously and describes its solution along with the forward kinematics and inverse kinematics of the mobile robot. The configuration of the pipe-climbing robot, including its hardware and software, is described, and the robot control with visual grasping is explained. The robot can be used practically for spray-pipe inspection as well as many potential other applications, such as inspection of the roof frame of a stadium consisting of pipes.

Seismic Test Result of Motor-Operated Butterfly Valve Actuators for Nuclear Power Plant

2018 ◽  
Author(s):  
Nobuo Kojima ◽  
Yoshitaka Tsutsumi ◽  
Kazuyoshi Yonekura ◽  
Koji Nishino ◽  
Yukio Watanabe ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Earthquake Resistance ◽  
Test Results ◽  
Butterfly Valve ◽  
Resistance Evaluation ◽  
Vibration Tests ◽  
Seismic Tests ◽  
Test Result

Seismic tests were carried out to confirm the operability limit acceleration for a standard motor-operated butterfly valve actuator in three excitation directions. Based on the results, seismic brackets that can operate valves even in three directions at 20 × 9.8 m/s2 or more were designed in three representative models. For the model subjected to the seismic test, we mounted the designed seismic bracket and confirmed the operability of the butterfly valve actuator by carrying out vibration tests in three directions at 20 × 9.8 m/s2. We used these results and previously reported motor-operated valve actuator seismic test results to creat a revision (draft) of the earthquake resistance evaluation methods.

scholarly journals Effects of %FIMA on Storage-Safety Parameters of Spent Fuel from Experimental Pebble-Bed Reactor

2021 ◽  
Vol 50 (2) ◽  
pp. 525-536
Author(s):  
Aisyah Aisyah ◽  
Mirawaty Mirawaty ◽  
Dwi Luhur Ibnu Saputra ◽  
Risdiyana Setiawan ◽  
Pungky Ayu Artian ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Pebble Bed ◽  
Energy Agency ◽  
Safety Level ◽  
Pebble Bed Reactor ◽  
Decay Heat

The back end of the utilization of nuclear technology is safety and management of spent fuel, which is a key element contributing to the success of the nuclear power plant program. Indonesia’s National Nuclear Energy Agency resolved to establish an experimental power reactor, called RDE, as a nuclear power plant demo. The fuel of this reactor is similar to that of German’s experimental pebble-bed reactor (PBR), Arbeitsgemeinschaft Versuchsreaktor(AVR). In this study, the spent fuel of AVR was studied to obtain the safety parameter data for storage of RDE spent fuel by varying the fission in the initial metallic atoms (%FIMA). These parameters that must be studied include the radioactivity, decay heat, proliferation threats of both 239Pu and 235U, and the presence of 137Cs, a dangerous fission product that can escape from damaged spent fuels. The calculation was conducted by ORIGEN 2.1. The result of the study demonstrates a higher %FIMA indicates a higher safety level that is required since the activity and decay heat of the spent fuel will increase and, as will be the total amounts of 239Pu and 137Cs. However, the 235U amount will decrease. For a 100 years storage of spent fuel, the optimum %FIMA is 8.2 with a canister capacity of 1,900 pebbles. Further, the activity and decay heat of the spent nuclear fuel are 2.013 × 1013 Bq and 6.065 W, respectively. The activities of 239Pu, 137Cs, and 235U are 5.187 ×1011, 7.100 × 1012, and 7.339 × 107 Bq, respectively.

Seismic Test Result of Air Operated Valve Actuators for Nuclear Power Plant (Air Operated Globe Valve (Diaphragm Type) and Air Operated Butterfly Valve (Lever Type) )

2021 ◽  
Author(s):  
Yoshitaka Tsutsumi ◽  
motoaki fuji ◽  
Yoshitsugu Nekomoto ◽  
Shigeki Suzuki ◽  
Yoshinao Matsubara ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Butterfly Valve ◽  
Globe Valve ◽  
Test Result

scholarly journals Kärnkraftverk som minnesplatser

1970 ◽  
pp. 96
Author(s):  
Anna Storm
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Cultural History ◽  
Nuclear Weapons ◽  
Nuclear Power ◽  
Power Plants ◽  
History Museum ◽  
The Past ◽  
Nuclear Plants ◽  
Museum Exhibitions

Is it possible to imagine a nuclear power plant as a memory site? In many respects, the answer is yes. Nuclear power already forms a part of cultural history museum exhibitions and closed-down nuclear plants have been documented and analysed from a heritage perspective. However, they are also highly controversial places for remembering the past, since their significance is exceptionally ambiguous. The plants stand for potential catastrophe, the complicated management of the radioactive waste and are often associated with nuclear weapons but – at the same time – they also imply hope in the future and a possible answer to the climate threat. Are closed-down nuclear power plants to be preserved in some way, and in that case why? And is our understanding of the concept of heritage to change in some respects as a consequence of this?

Research on Fault Diagnosis Approach Based on the Signed Directed Graph Model for Nuclear Power Plant

2013 ◽  
Author(s):  
Zhen Liu ◽  
Yong-kuo Liu ◽  
Xiao-tian Wu ◽  
Xian Zeng
Keyword(s):  
Fault Diagnosis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Directed Graph ◽  
Nuclear Power ◽  
Large Scale ◽  
Stable Operation ◽  
Strong Completeness ◽  
Signed Directed Graph ◽  
Propagation Paths

Signed Directed Graph (SDG) is a kind of fault diagnosis method based on a qualitative model. SDG can not only efficiently express the interrelations among variables of complex systems and has strong completeness, flexible reasoning ways and effective reasoning algorithm, but also shows faults propagation paths and faults detailed explanations. In this paper, SDG is applied to fault diagnosis in nuclear power plant which is a large-scale and complex system, and diagnosis results are transformed into a series of expert rules. The case studies show this method can find faults propagation paths and fault source can be determined effectively and improve the efficiency of fault diagnosis. The application of SDG in nuclear power plant can provide operation support to operators, so as to guarantee the safe and stable operation of the system.

Description, Output and Development Prospects of a 750 C Helium Direct Cycle Nuclear Power Plant With a Single Turbomachine and Intermediate Cooling

1977 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. Jude ◽  
J. Chaboseau
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Electricity Generation ◽  
Nuclear Power ◽  
First Generation ◽  
Safety Level ◽  
Good Efficiency ◽  
Higher Power ◽  
Intermediate Cooling ◽  
Development Prospects

The concept of a helium cycle nuclear power plant with a single turbomachine located outside the P.C.R.V. in a leaktight basement is being studied. Main features of the system are pointed out, as are the favorable aspects it offers for a satisfactory design of the plant, a high safety level, a good efficiency for electricity generation associated with a dry cooling tower and heat recovery in good conditions for heating or industrial purposes. Above all, the high technological credibility due to the temperature of 750 C is worth being emphasized. In this paper, a 750 C/800 MWe power plant (2000 MWt) with one intercooler is presented as an example of a possible first generation of gas turbine HTGR. Development prospects for higher power levels, 1000/2000 MWe, are also investigated, at first with one but also with two intermediate coolers, taking into consideration the somewhat higher-helium temperature. The net plant efficiencies, for electricity generation only, range from 40 to 45 percent.

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